January Docket Nos.
89
January 17, 1983 Docket Nos. 50-313, 50-302, 50-346, 50-312, 50-289, 50-269/270/287 SUBJECT: SUMMARY OF MEETING WITH THE BABCOCK & WILCOX OWNERS GROUP (B&WOG) CONCERNING THE B&WOG PRESSURIZED THERMAL SHOCK (PTS) PROGRAM January 11, 1983 Introduction The meeting was held in Bethesda, Maryland on January 11, 1983, at the request of the R&WOG to discuss the subject of the agenda Enclosure 1. The meeting was a followup to the meeting of December 20, 1982, with Mr. Denton to discuss the B&WOG Pressurized Thennal Shock Program and the applicability of the staff's screening criterion to B&W reactor plants. The attendees of the meeting are identifed in Enclosure 2. The material for the B&WOG presentation is included in Enclosure 3. Discussion The B&WOG indicated that several design features of B&W plants reduces the sensitivity of R&W plants to PTS events. These features included the operation of the vent valves during a transient, the relatively low inventory of the steam-generators and. the low vessel fluences. The analysis by the COMEX code shows that the vent valve fluid mixes with the HPI in the cold leg to heat the water into the down comer from 50 to 240 during a small break LOCA event. The staff expressed the concern that this condition has not been verified by experiment or test. The small inventory of the OTSG results in a fast acting blow dovin during a main steam line break (MSLB) event. Thus, the severity of MSLB event is less for the B&W plant than with the same event for a plant with U tube steam generators. The vessel fluences are lower because of a large gap between the core and vessel wall. After submitting RAW 1648 in response to NUREG-0737 II.K.2.13 the B&WOG decided to submit more realistic plant specific PTS analyses. Oconee 1 was the first which provided the ground work and the first probabilistic approach. The TMI-1 report was similar to the Oconee 1 report with more updated discussion concerning the mixing assumptions. The SBLOCA in the region of the size of the code safety valve (.023 sq. ft.) was determined to be the most severe (for PTS concern) transient. This would minimize the vent valve flow and natural circulation would he lost at 8-10'min. into the transient. However, natural circulation within the vessel would be maintained. The operator action would trip the RCP (immediately) and throttle HPI flow at,100 subcooling (93 min. into the transient). The failure (fail open) of the turbine bypass valves (both OTSGs) was determined to be the most severe overcooling transient. -_ OF I E I............ ......................... ........................ .............. **.....*....****************.... ............... SURNAME .... 9301250689 830117 " .*** . .... *....***.******.**. PDR ADOCK 05000269 DATE* p PDR . . NRC FORM 318 (1080) NRCM 0240 OFFICIAL RECORD COPY USso: 1s81--33s-96
Transcript of January Docket Nos.
Summary of 830111 meeting w/B&W Owners Group in Bethesda,MD re
Pressurized Thermal Shock Program.Docket Nos. 50-313, 50-302,
50-346, 50-312, 50-289, 50-269/270/287
SUBJECT: SUMMARY OF MEETING WITH THE BABCOCK & WILCOX OWNERS GROUP (B&WOG) CONCERNING THE B&WOG PRESSURIZED THERMAL SHOCK (PTS) PROGRAM January 11, 1983
Introduction
The meeting was held in Bethesda, Maryland on January 11, 1983, at the request of the R&WOG to discuss the subject of the agenda Enclosure 1. The meeting was a followup to the meeting of December 20, 1982, with Mr. Denton to discuss the B&WOG Pressurized Thennal Shock Program and the applicability of the staff's screening criterion to B&W reactor plants. The attendees of the meeting are identifed in Enclosure 2. The material for the B&WOG presentation is included in Enclosure 3.
Discussion
The B&WOG indicated that several design features of B&W plants reduces the sensitivity of R&W plants to PTS events. These features included the operation of the vent valves during a transient, the relatively low inventory of the steam-generators and. the low vessel fluences.
The analysis by the COMEX code shows that the vent valve fluid mixes with the HPI in the cold leg to heat the water into the down comer from 50 to 240 during a small break LOCA event. The staff expressed the concern that this condition has not been verified by experiment or test.
The small inventory of the OTSG results in a fast acting blow dovin during a main steam line break (MSLB) event. Thus, the severity of MSLB event is less for the B&W plant than with the same event for a plant with U tube steam generators. The vessel fluences are lower because of a large gap between the core and vessel wall.
After submitting RAW 1648 in response to NUREG-0737 II.K.2.13 the B&WOG decided to submit more realistic plant specific PTS analyses. Oconee 1 was the first which provided the ground work and the first probabilistic approach. The TMI-1 report was similar to the Oconee 1 report with more updated discussion concerning the mixing assumptions. The SBLOCA in the region of the size of the code safety valve (.023 sq. ft.) was determined to be the most severe (for PTS concern) transient. This would minimize the vent valve flow and natural circulation would he lost at 8-10'min. into the transient. However, natural circulation within the vessel would be maintained. The operator action would trip the RCP (immediately) and throttle HPI flow at,100 subcooling (93 min. into the transient). The failure (fail open) of the turbine bypass valves (both OTSGs) was determined to be the most severe overcooling transient.
-_ OF I E I............ ......................... ........................ .............. **.....*....****************.... ...............
DATE* p PDR . .
NRC FORM 318 (1080) NRCM 0240 OFFICIAL RECORD COPY USso: 1s81--33s-96
B&W Meeting Summary -2
Plant modification which have been or will be made and which reduces the sensitivity of B&W plants to PTS events include improved ICS/NNI power system reliability (as a result of the 1978 Rancho Seco "light bulb" event and the 1980 CR-3 event), upgraded auxiliary feedwater systems, anticipatory trips, inproved PORV reliability and improved system to reduce overfeed and excess heat removal transient. As a requirement of orders resulting from TMI-2 action, a failure made and effects analyses of each plant ICS has been provided the staff.
As a result of the B&WOG Integrated Reactor Vessel Material Surveillance Program initial fluence predictions for B&W vessels has decreased approximately 40%. Further decreases are expected for the low leakage fuel cycles.
The B&WOG concludes that there is no statistical difference between B&W operating plant events and the total industry events. Therefore the screening criterion should be valid for B&W plants. The original Oconee 1 report did not relate the PRA studies to the cummulative frequency versus T . New data presented in the meeting and in the Duke Power Company letter dated December 23, 1982 indicate good agreement with the staff's studies. Therefore, Duke concludes that the screening criterion is valid for B&W plants. However, the staff doesn't understand what was done to develop the T vs. frequence conclusions.
Conclusion
The B&WOG conclude that the B&W plants operating experience, plant design features, the plant specific evaluations and the Oconee probabilisitic analyses support the contention that the screening criteria is valid for B&W plants. The staff considered additional meetings are necessary to give understanding of the probabilistic work, of the mixing-phenomen and data concerning the vent valves and the overall completeness of the PTS evaluations related to the B&W plants. The staff would establish priorities and schedules for the additional meetings to resolve the above concerns.
riina signed by.
Guy Vissing, Project Manager Operating Reactors Branch #-4 Division of Licensing
Enclosures: 1. Agenda 2. Attendee List 3. Presentation
OFFICEO .. R , ... .. ....... ....... .. ... ........ . .... .... ... ..... .. ........ ..R. .p....... ....... F~G ng;cf
SURNAMEb ... 4 ...
1/./.83 .......................................................... DATE) _ _ _ _ _ _ _ _ _ _ _ _ 1_ _ _ _ _ _
NRC FORM 318 (10-80) NRCM 0240 OFFICIAL RECORD COPY.- USGPO: 1981-335-960
ORB#4:DL
Licensee:
* Copies also sent to those people on service (cc) list for subject plant(s).
Docket File NRC PDR L PDR ORB#4 Rdg GLainas JStolz Project Manager -GVissing Licensing Assistant-RIngram OELD Heltemes, AEOD IE SShowe (PWR) or CThayer (BWR), IE Meeting Summary File-ORB#4 RFraiey, ACRS-10 Program Support Branch
ORAB, Rm. 542 BGrimes, DEP SSchwartz, DEP SRamos, EPDB FPagano, EPLB
Meeting Participants Fm. NRC: CBerlinger KKniel WJohnston RRantala FSchroeder RWoods DBasdekas LLois DLieno JStrosnider CJohnson EThrom RKlecker JClifford SPawlicki
THERMAL SHOCK ISSUE DISTRIBUTION
R. Vollmer
W. Hazelton R. Mattson T. Speis T,. Murley H. Thompson
L. Shao R. Bernero AEOD E. Igne G. Knighton J. Austin J. Buzy J. Milhoan M. Vagins D. Ziemann
D. Garner R. Johnson E. Goodwin T. Novak J. Clifford
N. Randall
S. Chesnut A. Rubin C. Morris C. Serpan L.. Shotkin
A. Spano M. Virgilio T. Dunning C. Rossi
S. J. Bhatt R. Senseney A. Thadani S. Isreal NRC PDR Felix Litton
Enclosure 2
NAME ORGANIZATION
C. Berlinger NRC/CPB K. Kniel NRC/GIB E. Davidson FPC W. Johnston NRC/DE G. Lainas NRC/DL R. Rantala NRC/MTEB F. Schroeder NRC/DST R. Woods NRC/DST G. Vissing NRC/DL P. Tremblay ACRS T. Myers TECo J. Bohart B&W B. Short B&W P. Abraham Duke Power L., Gibson Consumers Power Co. D. Howard A P&L M. Snow AP&L H. Feinroth Gencon Corp. D. Basdekas NRC/RES L. Lois NRC/CPB D. Lieno NRC/CPB C. Whitmarsh B&W C. Hudson B&W T. Cogburn SMUD R. Gradomski TECo M. Foust TECo C. Hendrix Duke A. Lowe B&W J. Strosnider NRC/RES C. Johnson NRC D. Spond AP&L H. Slager Consumers Power Co. J. Pegram B&W K. Yoon B&W E. Throm NRC/DSI E. Wallace GPUN R. Gill Duke A. Rochino GPUN J. Delezenski GPUN R. Klecker NRC/MTEB J. Clifford, NRC/PSRB S. Pawlicki NRC/QE L. Conner NRC Calendar R. Ganthner B&W F. Walters B &W
Enclosure 1
NRC STAFF ON
PRESSURIZED THERMAL SHOCK
JANUARY 11, 1983
e B&W OG DATA SUPPORTING NRC SCREENING CRITERIA
- EVALUATION OF B&W OPERATING FRANK WALTERS
EXPERIENCE
JERRY DELEZENSKI
* SUMMARY
* DISCUSSION
NRC STAFF ON
PRESSURIZED THERMAL SHOCK
JANUARY 11, 1983
# DESIGN FEATURES OF B&W PLANTS FRANK WALTERS
* B&W OG DATA SUPPORTING NRC SCREENING CRITERIA
- EVALUATION OF B&W OPERATING FRANK WALTERS
EXPERIENCE
JERRY DELEZENSKI
e B&W OG PTS PROGRAM FRANK WALTERS LEE ROCHINO
e SUMMARY
* UPDATE NRC STAFF AND MANAGEMENT OF B&W OG
APPROACH TO PTS,
B&W OG ON PTS.
q B&W OG POSITION SUMMARY
- SCREENING CRITERIA IS CONSERVATIVELY VALID
- B&W OG HAS PROGRAMS THAT ASSURE THE CONTINUED
INTEGRITY OF THE BW RV AND ARE RESPONSIVE TO
STAFF'S CONCERNS
e MATERIAL PROPERTY DEFINITIONS
LATE * 1960's VERIFICATION OF EFFECT OF COPPER,
PHOSPHOROUS ON WELD TOUGHNESS.
* 1973 10 CFR 50, APPENDICES G AND H; CV USE
50 FT-LB LIMIT ESTABLISHED.
* 1974 SURVEILLANCE PROGRAMS IMPLEMENTED.
* 1976 INITIAL DISCUSSIONS WITHIN B&W AND OWNERS ABOUT REACTOR VESSEL MATERIALS.
* 1977 IMPLEMENTATION OF B8 OWNERS REACTOR
VESSEL MATERIALS PROGRAM.
BY STAFF TO EVALUATE'THERMAL SHOCK
CONCERN WITH SB LOCA AND EXTENDED HPI COOLING (B&W NSSS ONLY).
NRC REVISED IEB 79-05 RELFECTING COMMENTS BY UTILITIES ABOUT NEED FOR HPI FLOW AND
CONCERN FOR RV INTEGRITY.
'INITIAL SB LOCA OPERATOR-GUIDELINES CON TAIN GUIDANCE REFLECTING CONCERN FOR
REACTOR VESSEL INTEGRITY.
*1980 MAY MEETING BETWEFN B&W OWNERS AND NRC STAFF
TO DISCUSS PRELIMINARY RESULTS OF
BOUNDING CALCULATIONS.
AS WELL AS ADEQUATE CORE COOLING.
SEPT-Nov B&W OWNERS GRUOP INITIATED DISCUSSIONS WITH
CREARE IN ATTEMPT TO DEVELOP MIXING DATA
AND MIXING CODES.
WESTINGHOUSE REQUIRED BY 01/01/82.
PLAN ITEM II.K,2,13, AS A RESULT OF THIS
ANALYSIS, OWNERS GROUP DETERMINED THAT
GENERIC ANALYSES WERE OVERLY CONSERVATIVE
-- PLANS INITIATED FOR PLANT SPECIFIC
EVALUATIONS IN A TIMELY MANNER,
MARCH .INITIAL NRC/INDUSTRY MEETINGS REGARDING
PRESSURIZED THERMAL SHOCK.
PROGRAM SUBMITTED (PURSUANT TO 10 CFR 50,
APPENDIX G).
GROUP ACTIVITIES.
RELATIVE TO PTS--STAFF SEEKS GENERIC
LIMITS--OWNERS PLAN PLANT SPECIFIC
AUGUST OCONEE 1 REACTOR VESSEL 10-YEAR ISI UTILIZING R.G. 1.150.
SUMMER-FALL INDUSTRY EFFORT TO DEVELOP MIXING DATA
AND MIXING CODES.
RESOLVE PTS ISSUE; PLANT SPECIFIC SUB
MITTALS FOR LEAD PLANTS; FOLLOWED BY
PLANT SPECIFIC SUBMITTALS FOR REMAINING
PLANTS.
.ORNL PHASE 11 EFFORT STARTS.
JAN-FEB. OCONEE 2 REACTOR VESSEL 10-YEAR ISIs
MARCH MEETING WITH NRC STAFF TO DISCUSS OCONEE 1 SUBMITTAL. STAFF NOTES OCONEE PROBABILISTIC
ANALYSIS ON PTS WAS FIRST OF A KIND.
APRIL SUPPLMENTAL INFORMATION ON TRANSIENT
RESPONSES WITH DELAYED OPERATOR ACTIONS
PROVIDED,
*TMI-1 PLANT SPECIFIC EVALUATION SUBMITTED.
NOVEMBER ANO-1 REACTOR VESSEL 10-YEAR ISI,
DESIGN FEATURES OF B&W: PLANTS
LOCATION OF INTERNAL VENT VALVES
HEATING OF HPI FLUID IN COLD LEG
OTSG SECONDARY SIDE INVENTORY
LOW VESSEL FLUENCES
RCS Flow During Total Loss of Feedwater Event With a Small Break in Pressurizer
VALVE
VENT VALVE FLOW . RCSEfLOW ASSUMEO TO BE ZERO
'IIt
VO VESSEL DOINCOMER
3240F VENT VALVE
TEMPERATURE =50.F VESSEL 00 NCOMER
266.F
302.F
OUTLET NOZZLE VESSEL
00 SA- IS2 *l WF-8 WELDWELD
.93.9 .0 .92
TMI-I, II 225 400 209 >360
CR-3 250 > 400 250 305
RANCHO SECO 285 285 295 295
NRC DATA TAKEN FROM TABLE 3,1
REASONS FOR CHANGES IN CHARACTERIZATION OF TMI-II TFINAL
0 COLD LEG RTD, BECAUSE OF LOCATION PROVIDES
IMPROPER INFORMATION
- RTD PROVIDES USEFUL INFORMATION DURING FORCED
FLOW ONLY
VESSEL DOWNCOMER AND COLD LEG BECAUSE OF VENT
VALVE OPENING
400-500aF
FROM SECONDARY
0 POSSIBLE TEMPERATURE AS LOW As 250OF FOR 10-15
MINUTES WHEN ONE LOOP N/C WAS LOST
0 VESSEL NATURAL CIRCULATION CONTINUED THROUGH VENT
VALVE FLOW PATH
0 TCOLD READ 510aF BEFORE LOSS OF N/C AND 49 0 aF
IMMEDIATELY AFTER N/C RE-ESTABLISHED
0 CORE OUTLET TFMPERATURE WAS 500 0 F DURING ENTIRE PERIOD OF LOSS OF ONE LOOP NATURAL CIRCULATION
CONCLUSIONS FROM B&W PLANT OPERATING EXPERIENCE
0 BASED ON FREQUENCY OF OCCURRENCE THERE IS NO
DIFFERENCE BETWEEN B&W OPERATING EXPERIENCE AND
THE REST OF THE INDUSTRY.
O THUS, NRC RECOMMENDED SCREENING CRITERIA IS
CONSERVATIVE FOR B&W PLANTS
0 SEVERITY OF PTS PRECURSOR TRANSIENTS HAS BEEN
REDUCED SINCE CR-3 EVENT BECAUSE OF OPERATOR
TRAINING AND PLANT MODIFICATIONS
OPHR ATI NG EXPER IENCES (90% CONFIDENCE INTERVALS) ClACK EXTENSION WITHOUT ARREST
E1f B&W CONFIDENCE INTERVAL 1201 - CC W CONFIDENCE INTERVAL ED CE CONFIDENCE INTERVAL
OD 11n- E=ALL CONPIDENCE INTERVAL DM B&W MEAN FREQUENCY * W MEAN FR UENCY 100 - A CE MEAN FRE ENCY IV "ALL"MEAN FREQUENCY
930
70
60o
30
200 225 250 275 300 325 350 375 400 CRITICAL RTNDT (F)
e4 -7
500
400
0 15 30 45 60 75 90 105 3.5 HR
TimeMin.
Coto
rnl
Nyrn
PERFORMED AS PART OF 150 DAY REPORT (DPC-RS-1001)
TO COMPLEMENT THE DETERMINISTIC ANALYSIS
ANALYSIS DOCUMENTED IN SECTION 9 OF REPORT
REPORT SUBMITTED TO NRC ON 1-15-82
'DISCUSSED WITH NRC STAFF AT THE 3-24-82 MEETING
PURPOSE OF PTS PRA:
ALL POSSIBLE PTS ACCIDENT SEQUENCES WHICH
POTENTIALLY COMPROMISE RV INTEGRITY
CONSEQUENCES INTO CLASSES -SBLOCA, AND NON-LOCA
OVERCOOLING EVENTS
FAILURES, AND OPERATOR ERRORS NECESSARY TO
ACHIEVE SEVERE PTS CONDITIONS
* INITIATING EVENTS, SYSTEM LOGIC MODELS, AND
BASIC EVENT.DATA DERIVED FROM OCONEE PRA
PROGRAM
FEATURES:
INDUCED SMALL BREAK LOCA 'S
* CONSIDERED OVERCOOLING TRANSIENTS RESULTING
FAILURES, AND EXCESSIVE FEEDWATER
EVENTS CALCULATED TO BE-5.7 x 10-4/RY
(MEAN VALUE)
OTHER FEATURES:
,,SEQUENCES WHOSE TRANSIENT RESPONSE
WAS ANALYZED IN DETAIL
TEMPERATURE RANGES AND FREQUENCIES
FINAL FREQUENCY CUMULATIVE FREQUENCY
VESSEL FLUID (MEAN VALUE (MEAN VALUE TEMPERATURE, OF PER R-Y) PER R-Y)
200-300 5.1 x 1-4 5.1 x10
300-350 2,4 x 10- 2,9 x 10
350-U00 1.3 x 10-2 1,6 x 10-2
400-450 1.6 x 102 3.2 10-2
FREQUENCY BASED ON PRA STUDIES FINAL FLUID TEMPERATURE
LEGEND NRC STAFF PRA WEST INGHOUSE PRA
CL
10
5o oo0 150 200 75o 3oo 35o 4oo 450o 5oo
TE.MPERATURE .(DEG F)
RESULTS CONFIRM THAT SCREENING
LINEAR ALLOWABLE.
TRANSeANT ANALYSES HEAT TRANSFER/ VESSEL BAATIMALS FRACTURE MECHANICS
MIXING
ANALYSIS HIM FLOW Ws "xIOG
I VESSEL WALL THERMAL GRADIENT
LINEAR a
EVENT TEMPERATURE I0)
LEFM RESULTS FOR THERMAL SHOCK ANALYSES OF
OCONEE 1 REACTOR VESSEL
SMUD EVENT SA-1430(LW) 1.09 E 19 25
SBLOCA SA-1229(CW) 9.35 E 18 32 (WPS) SA-1585(CW) 1.23 E 19 32 (WPS)
OVERCOOLING SA-1430(LW) 1.09 E 19 25 (CASE 9)
Figure 2.1-1
o2 s
1. With Retor Coont Pu mps off operate in
Rgin II only.
I 1 or II1. REGION I EGION II
3. Wth Reactor Cootenq
aefe (W) highest incor
4. Mano nin the Reoo -10 Cooonet 509F auIso oed mka p nw e over the
Srite Frggatr Umot
L Wh Rerter Coolent 1000- ^wnvs off. the temer. .- 1800
atwuo oeat be kept within egonit. mal HPI ptew
1400 Wr
W o
a -, 1200
800 UNACCEPTABLE UNACCEPTABLE 00
REACTOR COOLANT TEMPERATURE (OF)
* PLANT SPECIFIC HPI
* CONSERVATIVE MIXING ASSUMPTIONS
* THIRTEEN CASES SELECTED FOR DETAILED THERMAL HYDRAULIC EVALUATION
* CASES CONSIDERED TO BE BOUNDING BASED ON ASSUMPTIONS INCORPORATED INTO EACH ANALYSIS
* UTILIZED PRESENT PLANT DESIGN AND OPERATOR PROCEDURES
* FREQUENCY OF PTS EVENTS
* FREQUENCY OF CASES CONSIDERED
SMALL BREAK LOCA
- HPI INITIATION/RCP TRIP/ESAS
SEVERITY
- WORST HOT LEG SIDE MECHANISTIC BREAK
MORE SEVERE THAN PORV LESS VENT VALVE FLOW THAN COLD LEG BREAK
MORE HPI ENTERS DOWNCOMER
MIN RCS UNIT DATE TEMP OF COMMENT
1 05/05/73 500 No OVERCOOLING RESULTED
2 01/04/74 422 COOLDOWN RATE 1400 F/HR
2 07/11/74 515 No OVERCOOLING RESULTED
2 09/10/74 NORMAL No OVERCOOLING RESULTED
2 09/17/74 547 No OVERCOOLING RESULTED
2 03/07/75 -540 No OVERCOOLING RESULTED
3 L/30/75 540 No OVERCOOLING RESULTED
3 05/25/75 485 No OVERCOOLING RESULTED
3 06/13/75 510 No OVERCOOLING RESULTED
3 07/13/75 NORMAL No OVERCOOLING RESULTED
1 08/14/76 NORMAL No OVERCOOLING RESULTED
1 12/14/78 500 No OVERCOOLING RESULTED
3 11/10/79 420 COOLDOWN RATE 1150F/HR
2 01/30/80 540 No OVERCOOLING RESULTED
3 03/14/80 546 No OVERCOOLING RESULTED
1 05/04/81 NORMAL No OVERCOOLING RESULTED
Figure 4.2-2 MIXING OCCURRING BELOW VESSEL INLET NOZZLE
VENT VALVES
Figure5.3-1 OCONEE I INSIDE SURFACE OF REACTOR VESSEL VELD LOCATIONS
x y z
11.V. FLANGE 223.5 REF. 36" I.0. MATING C.F. NOZZLE
SURFACE 28" 1.0.- I
C'4 56.9" SA -1135.
166.71911 SA-1073 c 433.992"
471. 105".
SA-1430 428.554"
536 4"
Figure 5.3-2 SB LOCA TRANSIENT WALL TEMPERATURE PROFILES FOR LOCATION OF SA-1493 WELD
600
550
INNER SURFACE
OF SA-1229 tELD
DNSIDE istance tnru vessel wa l, incnes OUTSIDE SURFACE SURFACE
Figure 5.3-4 CASE 9, WALL TEMPERATURE PROFILE VERSUS TIME FOR OVERCOOLING TRANSIENT
000
2. I 4.2 6.3 .
SURFACE
zI--
OF THE OCONEE UNIT 1 REACTOR VESSEL 32 EFPY
MATERIAL IDENTIFICATION(3) CHEMISTRY(1 ) NEUTRON INITI FLUENCE RD (Inside Weld or Heat Cu P Surface)
Number Type Location -vt. % wt. n/cm2 eF
AHR-54 SA508,CL2 Nozzle Belt 0.16 .006 1.97E18 (+60) Forging
SA-1135 circumfer- Nozzle Belt/ - - 1.97E18 (+20) ential weld Upper Shell
C2197 SA302B Upper Shell 0.15 0.008 9.35E18 (+40)
SA-1073 longitudi- Upper Shell - - 7.38E18 (+20) nal weld
SA-1229 circuafer- Upper Cir- - - 9.35E18 (+20) ential weld cumferential
(61% I.D.)
(1) Chemistry per BAW-1511P, October 1980. (Weld data is proprietary.)
(2) Estimated RTNDT values per BAW-10046A, Rev. 1, March 1976.
(3) Per BAW-1436, September 1977.
1-li
Figure 7.4-1 Location and Identification of aterials Used in Fabrication of Oconee Unit 1 Reactor Pressure Vessel
ZV2861 (Nozzle Belt) SA15261 Outlet
SA1494 Notzles Only
"N* SAl229 -61% (ID)
C3265-11 LOCATION Upper Shell
....... .. --- --- - -W112
OUTLET NOZZLE x VESSEL
190 SA-1430 WELD
220 SA-1O73 EELD
.70,. .7
OCONEE I REACTOR VESSEL
TRANSIENT WELD I.D. SMUD OVERCOOLJING CASE OR HEAT EVENT SBLOCA 9 T1 1 12
AHR 54 32 32 32 32 32 32
SA-1135 32 32 32 32 32 32
C2197-2 32 32 32 32 32 32
SA-1073 30 32 30 32 32 32
SA-1229 32 32(3) 32 32 32 32
SA-1493 30 32 31 32 32 32
C3278-1 32 32 32 32 32 32
SA-1585 32 32(3) 32 32 32 32
C2800-1 32 32 32 32 32 32
SA-1430 25 32 25 32 32 32
Vessel Inlet Nozzle 32 32 32 * NA
(1) Acceptance criteria is crack-arrest within 1/4T and no credit for
WPS, except where otherwise noted.
(2) Refer to Tables 6.6-1 and 6.6-2 for location.
(3) Results using war= prestressing (acceptance criteria is crack
arrest within 1/2T).
* Since Case 9 has a larger vessel temperature gradient and lower flaw
tip temperature, it is judged that Case 9 bounds Cases 10 and 11.
NA Not analyzeds
o TRANSIENTS EVALUATED ARE MORE SEVERE THAN
OPERATING EXPERIENCE
. EVALUATION OF SENSITIVITY OF OPERATOR RESPONSE
TIME SUBMITTED APRIL 30, 1982
* TEND TO SUPPORT VALIDITY OF STAFF'S SCREENING
CRITERIA
SUMMARY
CONSERVATIVE MIXING CALCULATIONS
PLANT SPECIFIC MATERIALS PROPERTIES
CONSERVATIVE POSTULATED FLAW SIZES
CONSERVATIVE CRACK ARREST CRITERIA
ESTIMATED FREQUENCY OF OCCURRENCE OF SEVERE REACTOR VESSEL THERMAL SHOCK EVENTS IS SMALL
EVOLUTJON nF THE
TMTJ-1 PTS ANALYSES
* DECISION MADE TO USE Til-1 REPORT AS NEXT STEP IN EVOLUTION OF B&W OG PTS PROGRAM USING:
1. INCREASE IN NUMBER OF TMI.-1 PLANT SPECIFIC PARAMETERS
2. TRANSIENT SELECTION CRITERIA
3. REALISTIC COMMIX IA MIxi Nt. PLj
a BENCHMARKED AGAINST AR FLOW TESTS
4. No WARM PRESTRESS, NO CRACK INITIATION
RESULT: 32 EFPY* RTNDT =35, F*
(CRITICAL LONG WELD)
J1IXING ANALYSES EMPLOYED IN
BW/OG THERMAL SHOCK ANALYSIS
PROVIDES CONSERVATIVE TEMPERATURE PROFILE AS SEEN IN COMPARISON
WITH CREARE DATA
FLOW 2-D CODE
ANALYSIS OF ACTUAL GEOMETRY AND SB LOCA TRANSIENT IN 2-D
GEOliETRIES PROVIDED TEMPERATURE PROFILES IN COLD LEG A D VESSEL
DOWJCOMER
- BENCHIMARKED AGAINST CREARE 1/5 SCALE MODEL DATA WITH GOOD PREDICTION
OF THE I IITED TEMPERATURE DATA
flIXING ANALYSES EMPLOYED IN
BW/OG THERMAL SHOCK ANALYSIS
TEMPERATURE PROFILES IN COLD LEG AND DOWNCOMER
- DETERMINED THAT SIGNIFICANT MIXING OCCURS IN COLD LEG
- QUALIFIED FOR.PTS EVALUATIONS VIA BENCHMARKS WITH
ANALYTICAL SOLUTIONS
- PRESENTLY BEING MODIFIED BY B&WAND INDUSTRY TO OBTAIN MORE
ACCURATE RESULTS
- JET TURBULENT iODEL - TOO CONSERVATIVE; AS SHOWN BY CREARE DATA
AND FLOW-2D AND COMMIX PREDICTIONS
- FLOW 2-D - GOOD AGREEMENT WITH CREARE DATA
- SUPPORTS COMMIX's RESULTS
- COMMIX-1A - PREDICTS RESULTS FOR BWI/ARC TEST DATA WITH GOOD
AGREEMENT
FLOW-2D AND COMMIX-lA PROVIDE REASONABLE TEMPERATURE PROFILES FOR
USE IN PTS ANALYSES
550 - --- - ""-"- 600 SEC.
200
0
0 X
to Turbulent Jet Analysis
7" UNDER NOZZLE FOR CREARE TEST DATA
27.5" UNDER NOZZLE FROM B&W ANALYSIS
.5 I8.8" UNDER NOZZLE FOR CREARE TEST DATA,
34.5" UNDER NOZZLE FROM B&W ANALYSIS
. 4
.3
.2
Distance From Nozzle < Incnes
Figure 4-14. Azimuthal Downcomer Temperature Comparison at 9.2 Inches Below Cold-Leg Centerline for CREARE Test No. 32
155
150
-- FLOW-20
125 I 8.0 10 12 14 16 18 20 22 24 26
Reference distance from not-leg center ine, inches
4-24 Babcock & Wilcox J,-
Figure 4-1. COMmixlA Computational Mesh for TI- Cold Leg and Vessel Downcomer
VENT VRC PUMP DISCHARGE
VENT
VALVES.
90
Figure 4-2. Location of TMI-1 Vessel Welds in.6, Z Coordinates for SBLOCA Transient
- ,,-VENT VALVES IN
PLANE 1=2
4-13
JCOLD LEG PIPNG
0.98
STATION 4 V,D STATION 1 2.8" 3.85" T ,D T
.STATION 5 1.84_ V, T VT 3.78" 0.406* REFERENCE WALL
6.81'
D - DYE INJECTION POINT
DATA
160
150
140
130
120 =%
Distance From Reference Wall (Vertical Leg), in.
4-18
W& NRC
SCREENING CRITERIA
GENERIC APPROACH
- SELECTION OF OVERCOOLING TRANSIENT
- VESSEL THERMAL ANALYSIS
- FRACTURE MECHANICS ANALYSIS
SB LOCA ACS TRANSIENT PRESSURE ANALYSIS HISTORY
OVERCOrCLING REPRESSURI11 ZATION FRACTURE
TRANSIENT RCS FLUID VESSEL MECHANIC VESSEL ANALYSIS PRESS/TEMP MIXING C WALL THERMAL ANALYSIS LIFE HIjS TOR Y ANALYSIS ANALYSI .S (EFPY).
VESSEL IlATERIAL PROPERTIES
BEGAN JULY 1981 - COMPLETED AUGUST 1982
* GENERIC APPROACH
- BAW-1628 (1980) ORIGINAL INDUSTRY T/S WORK
- BAW-1648 (1981)
PROVIDED DATA TO CONSTRUCT REASONABLE AND MECHANISTIC TRANSIENT
BREAK SIZE - LARGE BREAKS RESULT IN COMPLETE DEPRESSURIZING.
SiALL BREAKS RESULT IN HIGHER TEMPERATURES. MlOST PROBABLE
FAILURE IS PORVI HOWEVER, PZR CODE SAFETY VALVE, .023 FT.21 is
IN THE RANGE THAT RESULTS IN COLDEST DOWNCOMER TEMPERATURE 'WITHOUT COMPLETELY DEPRESSURIZING RCS.
BREAK LOCATION - BREAK ON TOP OF PRESSURIZER OR HOT LEGS. BREAK
IN COLD LEG IS LESS SEVERE BECAUSE: 1) HPI FLOWS OUT BREAK,
AND 2) INCREASED VENT VALVE FLOW THROUGH COLD LEG BREAK.
INITIATING EVENT - LOSS OF ALL FW - GIVE MECHANISTIC WAY OF
INITIATING TRANSIENT WHICH LIFT PZR CODE SAFETIES AND INTERRUPT
N/C.
Babcock&
(CONTINUED)
HPI FLOW - MAXIMUM HPI FLOW FOR OCONEE CLASS PLANTS IS 3 HPI PUMPS,
SG HEAT SINK AVAILABILITY - SINCE NATURAL CIRCULATION WAS NOT MAINTAINED, S.G. HEAT SINK IS NOT IMPORTANT,
OP ACTIONJ - TRIP RC PUMPS.
- THROTTLE HPI FLOW TO LIMIT CORE OUTLET SUBCOOLING
TO ~1000F (~-93 N INUTES).
Babcock &Wilco .'~~ti~~Jh.
OVERCOOLI:NG TRANSIENT
- CONSUMERS SENSITIVITY STUDY, DENTON's SHOW CAUSE LETTER
(50.54(8))
- SMUD TRANSIENT (3/20/78); LIGHT BULB INCIDENT
- RESPONSE TO I.E. BULLETIN 79-05C, EFFECTS OF TRIPPING
RC PUMPS ON NON-LOCA TRANSIENT
- SMUD FSAR - MSLB WITH TSV FAILURE IN OPPOSITE LOOP
OPERATING EXPERIENCE - OCONEE DPC SUBMITTAL TO NRC 1/15/82 (DPC-RS-1001)
- TMI-1 GPUN SUBMITTAL TO NRC 6/7/82
THIS INFORMATION PROVIDED GREATER THAN 60 INITIATING EVENTS OR COMB.
OF I.E./MULTI-FAILURES.
TRANSIENT.
- MSLB FOR FPC, AP&L, SMUD.
INITIATING EVENT - ICS SINGLE FAILURE - MECHAiNISTIC AND REALISTIC WAY
OF INITIATING TRANSIENT
Babcock &Wilcox
OVERCOOLING TRANSIENT
(CONTINUED)
AFW SYSTEM - INITIATES, FILLS AT vAX FLOW UNTIL LEVEL SETPOINT IS
REACHED: CONTROLS AT SETPOINT.
- NO THROTTLE OF HPI FOR MSLB TRANSIENT
- THROTTLE HPI AND SHUTOFF FOR TMI-1 TBV TRANSIENT
- 0 THROTTLE OF HPI FOR TBV 0-I TRANSIENT
Babcock Wilcor a Ucetmoott cmt-c.
OPERATING PROCEDURES AND TRAINING.
IN THE PROCESS OF-BEING IMPLEMENTED ON A SCHEDULE
CONSISTENT WITH REQUIREMENTS OF SECY 82-111
* CERTAIN ACTIONS HAVE ALREADY BEEN INCORPORATED
INTO EXISTING PROCEDURES.
* B&W OWNERS GROUP HAS BEEN WORKING CLOSELY WITH THE NRC STAFF IN THE REVIEW AND APPROVAL OF THE
GENERIC GUIDELINES
* GENERIC TRAINING MODULE ON PTS DEVELOPED AND IMPLEMENTED ON UTILITY SPECIFIC SCHEDULES
* NRC PTS AUDIT REVEALED NO SIGNIFICANT PROCEDURAL OR TRAINING DEFICIENCIES
PRESSURIZED THERMAL SHOCK TRAINING OUTLINE
REACTOR VESSEL THERMAL SHOCK DESCRIPTION
FACTORS AFFECTING REACTOR VESSEL THERMAL SHOCK
EFFECTS OF THESE FACTORS
PLANT MODIFICATIONS
COMPLETED/UNDERWAY WHICH
o IMPROVED ICS/NNI POWER SUPPLY RELIABILITY
(ADDRESSES 1978 R-S AND 1980 CR-3 EVENTS)
o UPGRADED Aux. FEEDWATER SYST.EMS
o ANTICIPATORY TRIPs, PORV RELIABILITY IMPROVED
0 IMPROVED SYSTEMS To REDUCE OVERFEED. AND
EXCESS HEAT REMOVAL TRANSIENTS
PHASE SHORT TERM PROGRAM
TAS KA- BEST ESTMIATE DESIGN CURVE.S TASK B REFINEMENT OF EUTRON FLUENCE TASK C - CHARACTERIZATION OF CHEMICAL COMPOSITION TASK D INTEGRATED SURVEI-LLANCE PROGRAM TASK E - COMPLIANCE WITH SECTION V, PARAGRAPH E.,
1OCFR50 APPENDIX G
PHASE II IRRADIATION OF TEST SPECIMENS
PHASE IV -EVALUATION OF IRRADIATED MATERIALS DATA
PHASE FRACTURE MECHANICS ANALYSIS
PHASE VIII - EVALUATION OF ATYPICAL WELD METAL
PHASE IX - DESIGN BASIS FLAW SIZE (ENHANCED ISI)
PHASE X -DOSIMETRY
Pr - I ~ -r~lo
PlI;SrS il-tv PP ASSSIII-IV
DEVELOP IIELD IETAL RESOLVE SitOtT FrA\CTURE TOUGMIESS TERM ISSUE DATA BASE .PHIASES, IX_ A X
RECUCE COurYATIS 'S 4IN DeSIGa CASES
DEVELOP ANALYSIS PROCEDURES
rrE VII
M41 I TOl /EVALUATE DMttSTRATE fttACTOR VESSEL R.Y. A J\ALING INTEGRITY BY ATU1LYSIS
PHASE VI
PAIMEAL ItcsuILi EMPHASIS ON ANALYSIS APPLICATIONS AND PPOCtiES PASED ON
STATE-OF-THE-ART DEVELOPMENTS IH INJDUSTRY-WIDE R&D PcXtPAPS.
BRP O\'ERS GROUP PTS PROGRAM
PER ENCLOSURE A (CHAPTER 9) OF SECY-82-465:
ANALYSIS/ACTIONS REQUIRED WHEN RTNDT SCREENING
CRITERIA ARE EXCEEDED OR WILL EXCEED WITHIN 3 YRS
A. VESSEL MATERIAL PROPERTIES
o REFINEMENT OF CHEMISTRY INFO FOR CRITERIA
MATERIALS
C. FLUX REDUCTION PROGRAMS
ABOVE ARE ADDRESSED BY ONGOING B&W OG INTEGRATED RV
MATERIAL SURVEILLANCE PROGRAM'(IRVMSP)
VESSEL MATERIAL PROPERTIES
MARCH 1976
BAW-1511P, OCTOBER 1980,
OCTOBER 1980, ALSO IN BAW-1485.
PHASE X - DOSIMETRY - METHODS TO FURTHER
REDUCE UNCERTAINTY
o OCONEE-I REPORT
o TMI-1 REPORT
o VESSEL WALL THICKNESS a CLAD THICKNESS: VESSEL INNER RADIUS
o LOCATION & ORIENTATION OF THE ASSUMED
INITIAL CRACK
PROPERTIES, K, E vs. TEMPERATURE
ASSUMED CRACK SHAPE AT INITIATION 8 TIME
OF INITIATION
o TREATMENT OF CLADDING INDUCED STRESSES
o UPPER SHELF TOUGHNESS
RTNDT (AT THE INNER VESSEL RADIUS)
NOTE:
RESULTS OF STAFF REVIEW OF TWO REPORTS SUBMITTED
Peak Reactor Vessel Fluence In Oconee 1
3.0
2.0 RV Surveillance Program
'onversion To Low Leakage Fuel Cycle 0 J
0 4 8 12 16 20 24 28 32
Effective Full Power Years ,(EFPY)
.Lq
' UY U u1KY K)U u U Lm UI Per NRC Procedure
AS OF 12/31/81- GUTHRIE + 2(T
32EFPY GUTHRIE + 2 O
CR-3
OCONEE
.I END
6 WITH ARIS (AUTOMATIC REACTOR INSPECTION SYSTEM), 0-I, O-II, o-III,
AND ANO-I HAVE COMPLETED THEIR 10-YEAR INSPECTION OF THE REACTOR
VESSELS.
* MEETINGS WITH THE NRC STAFF ESTABLISHED THAT THIS PLANNED INSPECTION
OF 0-I WOULD SATISFY THE INTENT OF RG 1.150 BEFORE THE TESTING WAS
BEGUN,
* THE NEAR SURFACE OF ALL BELTLINE WELDS WERE THOROUGHLY INSPECTED.
* WITH ARIS, IT WAS DEMONSTRATED WITH A HIGH DEGREE OF CONFIDENCE THAT
THE BELTLINE REGION OF THE VESSEL IS FREE OF SIGNIFICANT DEFECTS.
* THE INSPECTION RESULTS FOR 0-1, II, AND III ARE:
- NO INDICATION HAD A THRU-WALL DIMENSION GREATER THAN .15 INCHES.
- OF 133 INDICATIONS FOUND IN 0-1 VESSEL, THE VAST MAJORITY (114)
WERE CHARACTERIZED AS LAMINAR INDICATIONS, 3 AS PLANER SUBSURFACE
INDICATIONS AND 16 AS SLAG INCLUSIONS.
- 0-II HAD 4 INDICATIONS.
- 0-III HAD, 1 INDICATION.
- ANO-1 HAD 40 INDICATIONS (PRELIMINARY RESULTS - NOTHING SIGNIFICANT).
* ACCEPTANCE CRITERIA, FROM SECTION XI OF ASME CODE, IS THAT ALL
INDICATIONS BE LESS THAN 1/40 T OR SUPPLEMENTARY ANALYSIS BE PERFORMED
TO DEMONSTRATE THAT SUBSEQUENT OPERATION FOR THAT COMPONENT IS
JUSTIFIED.
TRIP TRANS ENT
PROGRAM TO FOLLOW THIS WORK
o ORNL -SUPPORTING ONGOING EVALUATION OF
0-1 BY ORN
MIXING 1/5 SCALE FACILITY ON B08 PLANTS
PROVIDED INFORMATION FOR CREARE
1/2 SCALE MODEL FACILITY
PER ENCLOSURE A (CHAPTER .9) OF SECY-82-,465:
ANALYSIS/ACTIONS REQUIRED WHEN RTNDT SCREENING
CRITERIA ARE EXCEEDED OR WILL EXCEED WITHIN 3 YRS
A. VESSEL MATERIAL PROPERTIES
o REFINEMENT OF CHEMISTRY INFO FOR CRITERIA
MATERIALS
C. FLUX REDUCTION PROGRAMIS
ABOVE ARE ADDRESSED BY ONGOING B&W OG INTEGRATED RV
MATERIAL SURVEILLANCE PROGRAM.(IRVMSP)
VESSEL MATERIAL PROPERTIES
MARCH 1976
BAW-1511P, OCTOBER 1980.
OCTOBER 1980, ALSO IN BAW-1485,
PHASE X - DOSIMETRY - METHODS TO FURTHER
REDUCE UNCERTAINTY
DETERMINISTIC FRACTURE MECHANICS
o OCONEE-I REPORT
o TMI-1 REPORT
VESSEL INNER RADIUS
PROPERTIES, K, Ed, vs. TEMPERATURE
o ASSUMED CRACK SHAPE AT INITIATION 9 TIME
OF IN ITIATION
o UPPER SHELF TOUGHNESS
RTNDT (AT THE INNER VESSEL RADIUS)
NOTE:
RESULTS OF STAFF REVIEW OF TWO REPORTS SUBMITTED
OTHER PROGRAMS SUPPORTED
BY ONERS GROUP
TRIP TRANSIENT
PROGRAM TO FOLLOW THIS WORK
o'. ORNL - SUPPORTING ONGOING EVALUATION OF
0-I BY ORNL
PROVIDED INFORMATION FOR CREARE
1/2 SCALE MODEL FACLITY
FUTURE ACTIONS SUPPORTING PTS
POSITION ON BW PLANTS
TRANSIENT SELECTION WAS CONSERVATIVE
PREDICTED TO REACH SCREENING CRITERIA
o IMPLEMENT PLANT.MODS AS NEEDED,TO RESPOND TO ANY
NEW ACTUAL PTS EVENTS
0 CONTINUE RV MATERIALS PROGRAM
o EVALUATE FURTHER FLUENCE REDUCTION MEASURES
O CONDUCT 10-YEAR INSERVICE INSPECTIONS OF REACTOR
VESSEL WELDS WITH ENHANCED ISM 1ETHODOLOGY
o CONTINUE TRANSIENT ASSESSMENT PROGRAM
o CONTINUE INVOLVEMENT AND SUPPORT OF INDUSTRY
PROGRAMS
PROGRAM IN PLACE WHICH IS
RESPONSIVE TO THE ITEMS IN
CHAPTER 9 OF STAFF'S REPORT.
o SCREENING CRITERIA IS
PLAN,!TS,
ESSENTIAL TO COMPLETE THE
DEMONSTRATION THAT PTS RISK
SUBJECT: SUMMARY OF MEETING WITH THE BABCOCK & WILCOX OWNERS GROUP (B&WOG) CONCERNING THE B&WOG PRESSURIZED THERMAL SHOCK (PTS) PROGRAM January 11, 1983
Introduction
The meeting was held in Bethesda, Maryland on January 11, 1983, at the request of the R&WOG to discuss the subject of the agenda Enclosure 1. The meeting was a followup to the meeting of December 20, 1982, with Mr. Denton to discuss the B&WOG Pressurized Thennal Shock Program and the applicability of the staff's screening criterion to B&W reactor plants. The attendees of the meeting are identifed in Enclosure 2. The material for the B&WOG presentation is included in Enclosure 3.
Discussion
The B&WOG indicated that several design features of B&W plants reduces the sensitivity of R&W plants to PTS events. These features included the operation of the vent valves during a transient, the relatively low inventory of the steam-generators and. the low vessel fluences.
The analysis by the COMEX code shows that the vent valve fluid mixes with the HPI in the cold leg to heat the water into the down comer from 50 to 240 during a small break LOCA event. The staff expressed the concern that this condition has not been verified by experiment or test.
The small inventory of the OTSG results in a fast acting blow dovin during a main steam line break (MSLB) event. Thus, the severity of MSLB event is less for the B&W plant than with the same event for a plant with U tube steam generators. The vessel fluences are lower because of a large gap between the core and vessel wall.
After submitting RAW 1648 in response to NUREG-0737 II.K.2.13 the B&WOG decided to submit more realistic plant specific PTS analyses. Oconee 1 was the first which provided the ground work and the first probabilistic approach. The TMI-1 report was similar to the Oconee 1 report with more updated discussion concerning the mixing assumptions. The SBLOCA in the region of the size of the code safety valve (.023 sq. ft.) was determined to be the most severe (for PTS concern) transient. This would minimize the vent valve flow and natural circulation would he lost at 8-10'min. into the transient. However, natural circulation within the vessel would be maintained. The operator action would trip the RCP (immediately) and throttle HPI flow at,100 subcooling (93 min. into the transient). The failure (fail open) of the turbine bypass valves (both OTSGs) was determined to be the most severe overcooling transient.
-_ OF I E I............ ......................... ........................ .............. **.....*....****************.... ...............
DATE* p PDR . .
NRC FORM 318 (1080) NRCM 0240 OFFICIAL RECORD COPY USso: 1s81--33s-96
B&W Meeting Summary -2
Plant modification which have been or will be made and which reduces the sensitivity of B&W plants to PTS events include improved ICS/NNI power system reliability (as a result of the 1978 Rancho Seco "light bulb" event and the 1980 CR-3 event), upgraded auxiliary feedwater systems, anticipatory trips, inproved PORV reliability and improved system to reduce overfeed and excess heat removal transient. As a requirement of orders resulting from TMI-2 action, a failure made and effects analyses of each plant ICS has been provided the staff.
As a result of the B&WOG Integrated Reactor Vessel Material Surveillance Program initial fluence predictions for B&W vessels has decreased approximately 40%. Further decreases are expected for the low leakage fuel cycles.
The B&WOG concludes that there is no statistical difference between B&W operating plant events and the total industry events. Therefore the screening criterion should be valid for B&W plants. The original Oconee 1 report did not relate the PRA studies to the cummulative frequency versus T . New data presented in the meeting and in the Duke Power Company letter dated December 23, 1982 indicate good agreement with the staff's studies. Therefore, Duke concludes that the screening criterion is valid for B&W plants. However, the staff doesn't understand what was done to develop the T vs. frequence conclusions.
Conclusion
The B&WOG conclude that the B&W plants operating experience, plant design features, the plant specific evaluations and the Oconee probabilisitic analyses support the contention that the screening criteria is valid for B&W plants. The staff considered additional meetings are necessary to give understanding of the probabilistic work, of the mixing-phenomen and data concerning the vent valves and the overall completeness of the PTS evaluations related to the B&W plants. The staff would establish priorities and schedules for the additional meetings to resolve the above concerns.
riina signed by.
Guy Vissing, Project Manager Operating Reactors Branch #-4 Division of Licensing
Enclosures: 1. Agenda 2. Attendee List 3. Presentation
OFFICEO .. R , ... .. ....... ....... .. ... ........ . .... .... ... ..... .. ........ ..R. .p....... ....... F~G ng;cf
SURNAMEb ... 4 ...
1/./.83 .......................................................... DATE) _ _ _ _ _ _ _ _ _ _ _ _ 1_ _ _ _ _ _
NRC FORM 318 (10-80) NRCM 0240 OFFICIAL RECORD COPY.- USGPO: 1981-335-960
ORB#4:DL
Licensee:
* Copies also sent to those people on service (cc) list for subject plant(s).
Docket File NRC PDR L PDR ORB#4 Rdg GLainas JStolz Project Manager -GVissing Licensing Assistant-RIngram OELD Heltemes, AEOD IE SShowe (PWR) or CThayer (BWR), IE Meeting Summary File-ORB#4 RFraiey, ACRS-10 Program Support Branch
ORAB, Rm. 542 BGrimes, DEP SSchwartz, DEP SRamos, EPDB FPagano, EPLB
Meeting Participants Fm. NRC: CBerlinger KKniel WJohnston RRantala FSchroeder RWoods DBasdekas LLois DLieno JStrosnider CJohnson EThrom RKlecker JClifford SPawlicki
THERMAL SHOCK ISSUE DISTRIBUTION
R. Vollmer
W. Hazelton R. Mattson T. Speis T,. Murley H. Thompson
L. Shao R. Bernero AEOD E. Igne G. Knighton J. Austin J. Buzy J. Milhoan M. Vagins D. Ziemann
D. Garner R. Johnson E. Goodwin T. Novak J. Clifford
N. Randall
S. Chesnut A. Rubin C. Morris C. Serpan L.. Shotkin
A. Spano M. Virgilio T. Dunning C. Rossi
S. J. Bhatt R. Senseney A. Thadani S. Isreal NRC PDR Felix Litton
Enclosure 2
NAME ORGANIZATION
C. Berlinger NRC/CPB K. Kniel NRC/GIB E. Davidson FPC W. Johnston NRC/DE G. Lainas NRC/DL R. Rantala NRC/MTEB F. Schroeder NRC/DST R. Woods NRC/DST G. Vissing NRC/DL P. Tremblay ACRS T. Myers TECo J. Bohart B&W B. Short B&W P. Abraham Duke Power L., Gibson Consumers Power Co. D. Howard A P&L M. Snow AP&L H. Feinroth Gencon Corp. D. Basdekas NRC/RES L. Lois NRC/CPB D. Lieno NRC/CPB C. Whitmarsh B&W C. Hudson B&W T. Cogburn SMUD R. Gradomski TECo M. Foust TECo C. Hendrix Duke A. Lowe B&W J. Strosnider NRC/RES C. Johnson NRC D. Spond AP&L H. Slager Consumers Power Co. J. Pegram B&W K. Yoon B&W E. Throm NRC/DSI E. Wallace GPUN R. Gill Duke A. Rochino GPUN J. Delezenski GPUN R. Klecker NRC/MTEB J. Clifford, NRC/PSRB S. Pawlicki NRC/QE L. Conner NRC Calendar R. Ganthner B&W F. Walters B &W
Enclosure 1
NRC STAFF ON
PRESSURIZED THERMAL SHOCK
JANUARY 11, 1983
e B&W OG DATA SUPPORTING NRC SCREENING CRITERIA
- EVALUATION OF B&W OPERATING FRANK WALTERS
EXPERIENCE
JERRY DELEZENSKI
* SUMMARY
* DISCUSSION
NRC STAFF ON
PRESSURIZED THERMAL SHOCK
JANUARY 11, 1983
# DESIGN FEATURES OF B&W PLANTS FRANK WALTERS
* B&W OG DATA SUPPORTING NRC SCREENING CRITERIA
- EVALUATION OF B&W OPERATING FRANK WALTERS
EXPERIENCE
JERRY DELEZENSKI
e B&W OG PTS PROGRAM FRANK WALTERS LEE ROCHINO
e SUMMARY
* UPDATE NRC STAFF AND MANAGEMENT OF B&W OG
APPROACH TO PTS,
B&W OG ON PTS.
q B&W OG POSITION SUMMARY
- SCREENING CRITERIA IS CONSERVATIVELY VALID
- B&W OG HAS PROGRAMS THAT ASSURE THE CONTINUED
INTEGRITY OF THE BW RV AND ARE RESPONSIVE TO
STAFF'S CONCERNS
e MATERIAL PROPERTY DEFINITIONS
LATE * 1960's VERIFICATION OF EFFECT OF COPPER,
PHOSPHOROUS ON WELD TOUGHNESS.
* 1973 10 CFR 50, APPENDICES G AND H; CV USE
50 FT-LB LIMIT ESTABLISHED.
* 1974 SURVEILLANCE PROGRAMS IMPLEMENTED.
* 1976 INITIAL DISCUSSIONS WITHIN B&W AND OWNERS ABOUT REACTOR VESSEL MATERIALS.
* 1977 IMPLEMENTATION OF B8 OWNERS REACTOR
VESSEL MATERIALS PROGRAM.
BY STAFF TO EVALUATE'THERMAL SHOCK
CONCERN WITH SB LOCA AND EXTENDED HPI COOLING (B&W NSSS ONLY).
NRC REVISED IEB 79-05 RELFECTING COMMENTS BY UTILITIES ABOUT NEED FOR HPI FLOW AND
CONCERN FOR RV INTEGRITY.
'INITIAL SB LOCA OPERATOR-GUIDELINES CON TAIN GUIDANCE REFLECTING CONCERN FOR
REACTOR VESSEL INTEGRITY.
*1980 MAY MEETING BETWEFN B&W OWNERS AND NRC STAFF
TO DISCUSS PRELIMINARY RESULTS OF
BOUNDING CALCULATIONS.
AS WELL AS ADEQUATE CORE COOLING.
SEPT-Nov B&W OWNERS GRUOP INITIATED DISCUSSIONS WITH
CREARE IN ATTEMPT TO DEVELOP MIXING DATA
AND MIXING CODES.
WESTINGHOUSE REQUIRED BY 01/01/82.
PLAN ITEM II.K,2,13, AS A RESULT OF THIS
ANALYSIS, OWNERS GROUP DETERMINED THAT
GENERIC ANALYSES WERE OVERLY CONSERVATIVE
-- PLANS INITIATED FOR PLANT SPECIFIC
EVALUATIONS IN A TIMELY MANNER,
MARCH .INITIAL NRC/INDUSTRY MEETINGS REGARDING
PRESSURIZED THERMAL SHOCK.
PROGRAM SUBMITTED (PURSUANT TO 10 CFR 50,
APPENDIX G).
GROUP ACTIVITIES.
RELATIVE TO PTS--STAFF SEEKS GENERIC
LIMITS--OWNERS PLAN PLANT SPECIFIC
AUGUST OCONEE 1 REACTOR VESSEL 10-YEAR ISI UTILIZING R.G. 1.150.
SUMMER-FALL INDUSTRY EFFORT TO DEVELOP MIXING DATA
AND MIXING CODES.
RESOLVE PTS ISSUE; PLANT SPECIFIC SUB
MITTALS FOR LEAD PLANTS; FOLLOWED BY
PLANT SPECIFIC SUBMITTALS FOR REMAINING
PLANTS.
.ORNL PHASE 11 EFFORT STARTS.
JAN-FEB. OCONEE 2 REACTOR VESSEL 10-YEAR ISIs
MARCH MEETING WITH NRC STAFF TO DISCUSS OCONEE 1 SUBMITTAL. STAFF NOTES OCONEE PROBABILISTIC
ANALYSIS ON PTS WAS FIRST OF A KIND.
APRIL SUPPLMENTAL INFORMATION ON TRANSIENT
RESPONSES WITH DELAYED OPERATOR ACTIONS
PROVIDED,
*TMI-1 PLANT SPECIFIC EVALUATION SUBMITTED.
NOVEMBER ANO-1 REACTOR VESSEL 10-YEAR ISI,
DESIGN FEATURES OF B&W: PLANTS
LOCATION OF INTERNAL VENT VALVES
HEATING OF HPI FLUID IN COLD LEG
OTSG SECONDARY SIDE INVENTORY
LOW VESSEL FLUENCES
RCS Flow During Total Loss of Feedwater Event With a Small Break in Pressurizer
VALVE
VENT VALVE FLOW . RCSEfLOW ASSUMEO TO BE ZERO
'IIt
VO VESSEL DOINCOMER
3240F VENT VALVE
TEMPERATURE =50.F VESSEL 00 NCOMER
266.F
302.F
OUTLET NOZZLE VESSEL
00 SA- IS2 *l WF-8 WELDWELD
.93.9 .0 .92
TMI-I, II 225 400 209 >360
CR-3 250 > 400 250 305
RANCHO SECO 285 285 295 295
NRC DATA TAKEN FROM TABLE 3,1
REASONS FOR CHANGES IN CHARACTERIZATION OF TMI-II TFINAL
0 COLD LEG RTD, BECAUSE OF LOCATION PROVIDES
IMPROPER INFORMATION
- RTD PROVIDES USEFUL INFORMATION DURING FORCED
FLOW ONLY
VESSEL DOWNCOMER AND COLD LEG BECAUSE OF VENT
VALVE OPENING
400-500aF
FROM SECONDARY
0 POSSIBLE TEMPERATURE AS LOW As 250OF FOR 10-15
MINUTES WHEN ONE LOOP N/C WAS LOST
0 VESSEL NATURAL CIRCULATION CONTINUED THROUGH VENT
VALVE FLOW PATH
0 TCOLD READ 510aF BEFORE LOSS OF N/C AND 49 0 aF
IMMEDIATELY AFTER N/C RE-ESTABLISHED
0 CORE OUTLET TFMPERATURE WAS 500 0 F DURING ENTIRE PERIOD OF LOSS OF ONE LOOP NATURAL CIRCULATION
CONCLUSIONS FROM B&W PLANT OPERATING EXPERIENCE
0 BASED ON FREQUENCY OF OCCURRENCE THERE IS NO
DIFFERENCE BETWEEN B&W OPERATING EXPERIENCE AND
THE REST OF THE INDUSTRY.
O THUS, NRC RECOMMENDED SCREENING CRITERIA IS
CONSERVATIVE FOR B&W PLANTS
0 SEVERITY OF PTS PRECURSOR TRANSIENTS HAS BEEN
REDUCED SINCE CR-3 EVENT BECAUSE OF OPERATOR
TRAINING AND PLANT MODIFICATIONS
OPHR ATI NG EXPER IENCES (90% CONFIDENCE INTERVALS) ClACK EXTENSION WITHOUT ARREST
E1f B&W CONFIDENCE INTERVAL 1201 - CC W CONFIDENCE INTERVAL ED CE CONFIDENCE INTERVAL
OD 11n- E=ALL CONPIDENCE INTERVAL DM B&W MEAN FREQUENCY * W MEAN FR UENCY 100 - A CE MEAN FRE ENCY IV "ALL"MEAN FREQUENCY
930
70
60o
30
200 225 250 275 300 325 350 375 400 CRITICAL RTNDT (F)
e4 -7
500
400
0 15 30 45 60 75 90 105 3.5 HR
TimeMin.
Coto
rnl
Nyrn
PERFORMED AS PART OF 150 DAY REPORT (DPC-RS-1001)
TO COMPLEMENT THE DETERMINISTIC ANALYSIS
ANALYSIS DOCUMENTED IN SECTION 9 OF REPORT
REPORT SUBMITTED TO NRC ON 1-15-82
'DISCUSSED WITH NRC STAFF AT THE 3-24-82 MEETING
PURPOSE OF PTS PRA:
ALL POSSIBLE PTS ACCIDENT SEQUENCES WHICH
POTENTIALLY COMPROMISE RV INTEGRITY
CONSEQUENCES INTO CLASSES -SBLOCA, AND NON-LOCA
OVERCOOLING EVENTS
FAILURES, AND OPERATOR ERRORS NECESSARY TO
ACHIEVE SEVERE PTS CONDITIONS
* INITIATING EVENTS, SYSTEM LOGIC MODELS, AND
BASIC EVENT.DATA DERIVED FROM OCONEE PRA
PROGRAM
FEATURES:
INDUCED SMALL BREAK LOCA 'S
* CONSIDERED OVERCOOLING TRANSIENTS RESULTING
FAILURES, AND EXCESSIVE FEEDWATER
EVENTS CALCULATED TO BE-5.7 x 10-4/RY
(MEAN VALUE)
OTHER FEATURES:
,,SEQUENCES WHOSE TRANSIENT RESPONSE
WAS ANALYZED IN DETAIL
TEMPERATURE RANGES AND FREQUENCIES
FINAL FREQUENCY CUMULATIVE FREQUENCY
VESSEL FLUID (MEAN VALUE (MEAN VALUE TEMPERATURE, OF PER R-Y) PER R-Y)
200-300 5.1 x 1-4 5.1 x10
300-350 2,4 x 10- 2,9 x 10
350-U00 1.3 x 10-2 1,6 x 10-2
400-450 1.6 x 102 3.2 10-2
FREQUENCY BASED ON PRA STUDIES FINAL FLUID TEMPERATURE
LEGEND NRC STAFF PRA WEST INGHOUSE PRA
CL
10
5o oo0 150 200 75o 3oo 35o 4oo 450o 5oo
TE.MPERATURE .(DEG F)
RESULTS CONFIRM THAT SCREENING
LINEAR ALLOWABLE.
TRANSeANT ANALYSES HEAT TRANSFER/ VESSEL BAATIMALS FRACTURE MECHANICS
MIXING
ANALYSIS HIM FLOW Ws "xIOG
I VESSEL WALL THERMAL GRADIENT
LINEAR a
EVENT TEMPERATURE I0)
LEFM RESULTS FOR THERMAL SHOCK ANALYSES OF
OCONEE 1 REACTOR VESSEL
SMUD EVENT SA-1430(LW) 1.09 E 19 25
SBLOCA SA-1229(CW) 9.35 E 18 32 (WPS) SA-1585(CW) 1.23 E 19 32 (WPS)
OVERCOOLING SA-1430(LW) 1.09 E 19 25 (CASE 9)
Figure 2.1-1
o2 s
1. With Retor Coont Pu mps off operate in
Rgin II only.
I 1 or II1. REGION I EGION II
3. Wth Reactor Cootenq
aefe (W) highest incor
4. Mano nin the Reoo -10 Cooonet 509F auIso oed mka p nw e over the
Srite Frggatr Umot
L Wh Rerter Coolent 1000- ^wnvs off. the temer. .- 1800
atwuo oeat be kept within egonit. mal HPI ptew
1400 Wr
W o
a -, 1200
800 UNACCEPTABLE UNACCEPTABLE 00
REACTOR COOLANT TEMPERATURE (OF)
* PLANT SPECIFIC HPI
* CONSERVATIVE MIXING ASSUMPTIONS
* THIRTEEN CASES SELECTED FOR DETAILED THERMAL HYDRAULIC EVALUATION
* CASES CONSIDERED TO BE BOUNDING BASED ON ASSUMPTIONS INCORPORATED INTO EACH ANALYSIS
* UTILIZED PRESENT PLANT DESIGN AND OPERATOR PROCEDURES
* FREQUENCY OF PTS EVENTS
* FREQUENCY OF CASES CONSIDERED
SMALL BREAK LOCA
- HPI INITIATION/RCP TRIP/ESAS
SEVERITY
- WORST HOT LEG SIDE MECHANISTIC BREAK
MORE SEVERE THAN PORV LESS VENT VALVE FLOW THAN COLD LEG BREAK
MORE HPI ENTERS DOWNCOMER
MIN RCS UNIT DATE TEMP OF COMMENT
1 05/05/73 500 No OVERCOOLING RESULTED
2 01/04/74 422 COOLDOWN RATE 1400 F/HR
2 07/11/74 515 No OVERCOOLING RESULTED
2 09/10/74 NORMAL No OVERCOOLING RESULTED
2 09/17/74 547 No OVERCOOLING RESULTED
2 03/07/75 -540 No OVERCOOLING RESULTED
3 L/30/75 540 No OVERCOOLING RESULTED
3 05/25/75 485 No OVERCOOLING RESULTED
3 06/13/75 510 No OVERCOOLING RESULTED
3 07/13/75 NORMAL No OVERCOOLING RESULTED
1 08/14/76 NORMAL No OVERCOOLING RESULTED
1 12/14/78 500 No OVERCOOLING RESULTED
3 11/10/79 420 COOLDOWN RATE 1150F/HR
2 01/30/80 540 No OVERCOOLING RESULTED
3 03/14/80 546 No OVERCOOLING RESULTED
1 05/04/81 NORMAL No OVERCOOLING RESULTED
Figure 4.2-2 MIXING OCCURRING BELOW VESSEL INLET NOZZLE
VENT VALVES
Figure5.3-1 OCONEE I INSIDE SURFACE OF REACTOR VESSEL VELD LOCATIONS
x y z
11.V. FLANGE 223.5 REF. 36" I.0. MATING C.F. NOZZLE
SURFACE 28" 1.0.- I
C'4 56.9" SA -1135.
166.71911 SA-1073 c 433.992"
471. 105".
SA-1430 428.554"
536 4"
Figure 5.3-2 SB LOCA TRANSIENT WALL TEMPERATURE PROFILES FOR LOCATION OF SA-1493 WELD
600
550
INNER SURFACE
OF SA-1229 tELD
DNSIDE istance tnru vessel wa l, incnes OUTSIDE SURFACE SURFACE
Figure 5.3-4 CASE 9, WALL TEMPERATURE PROFILE VERSUS TIME FOR OVERCOOLING TRANSIENT
000
2. I 4.2 6.3 .
SURFACE
zI--
OF THE OCONEE UNIT 1 REACTOR VESSEL 32 EFPY
MATERIAL IDENTIFICATION(3) CHEMISTRY(1 ) NEUTRON INITI FLUENCE RD (Inside Weld or Heat Cu P Surface)
Number Type Location -vt. % wt. n/cm2 eF
AHR-54 SA508,CL2 Nozzle Belt 0.16 .006 1.97E18 (+60) Forging
SA-1135 circumfer- Nozzle Belt/ - - 1.97E18 (+20) ential weld Upper Shell
C2197 SA302B Upper Shell 0.15 0.008 9.35E18 (+40)
SA-1073 longitudi- Upper Shell - - 7.38E18 (+20) nal weld
SA-1229 circuafer- Upper Cir- - - 9.35E18 (+20) ential weld cumferential
(61% I.D.)
(1) Chemistry per BAW-1511P, October 1980. (Weld data is proprietary.)
(2) Estimated RTNDT values per BAW-10046A, Rev. 1, March 1976.
(3) Per BAW-1436, September 1977.
1-li
Figure 7.4-1 Location and Identification of aterials Used in Fabrication of Oconee Unit 1 Reactor Pressure Vessel
ZV2861 (Nozzle Belt) SA15261 Outlet
SA1494 Notzles Only
"N* SAl229 -61% (ID)
C3265-11 LOCATION Upper Shell
....... .. --- --- - -W112
OUTLET NOZZLE x VESSEL
190 SA-1430 WELD
220 SA-1O73 EELD
.70,. .7
OCONEE I REACTOR VESSEL
TRANSIENT WELD I.D. SMUD OVERCOOLJING CASE OR HEAT EVENT SBLOCA 9 T1 1 12
AHR 54 32 32 32 32 32 32
SA-1135 32 32 32 32 32 32
C2197-2 32 32 32 32 32 32
SA-1073 30 32 30 32 32 32
SA-1229 32 32(3) 32 32 32 32
SA-1493 30 32 31 32 32 32
C3278-1 32 32 32 32 32 32
SA-1585 32 32(3) 32 32 32 32
C2800-1 32 32 32 32 32 32
SA-1430 25 32 25 32 32 32
Vessel Inlet Nozzle 32 32 32 * NA
(1) Acceptance criteria is crack-arrest within 1/4T and no credit for
WPS, except where otherwise noted.
(2) Refer to Tables 6.6-1 and 6.6-2 for location.
(3) Results using war= prestressing (acceptance criteria is crack
arrest within 1/2T).
* Since Case 9 has a larger vessel temperature gradient and lower flaw
tip temperature, it is judged that Case 9 bounds Cases 10 and 11.
NA Not analyzeds
o TRANSIENTS EVALUATED ARE MORE SEVERE THAN
OPERATING EXPERIENCE
. EVALUATION OF SENSITIVITY OF OPERATOR RESPONSE
TIME SUBMITTED APRIL 30, 1982
* TEND TO SUPPORT VALIDITY OF STAFF'S SCREENING
CRITERIA
SUMMARY
CONSERVATIVE MIXING CALCULATIONS
PLANT SPECIFIC MATERIALS PROPERTIES
CONSERVATIVE POSTULATED FLAW SIZES
CONSERVATIVE CRACK ARREST CRITERIA
ESTIMATED FREQUENCY OF OCCURRENCE OF SEVERE REACTOR VESSEL THERMAL SHOCK EVENTS IS SMALL
EVOLUTJON nF THE
TMTJ-1 PTS ANALYSES
* DECISION MADE TO USE Til-1 REPORT AS NEXT STEP IN EVOLUTION OF B&W OG PTS PROGRAM USING:
1. INCREASE IN NUMBER OF TMI.-1 PLANT SPECIFIC PARAMETERS
2. TRANSIENT SELECTION CRITERIA
3. REALISTIC COMMIX IA MIxi Nt. PLj
a BENCHMARKED AGAINST AR FLOW TESTS
4. No WARM PRESTRESS, NO CRACK INITIATION
RESULT: 32 EFPY* RTNDT =35, F*
(CRITICAL LONG WELD)
J1IXING ANALYSES EMPLOYED IN
BW/OG THERMAL SHOCK ANALYSIS
PROVIDES CONSERVATIVE TEMPERATURE PROFILE AS SEEN IN COMPARISON
WITH CREARE DATA
FLOW 2-D CODE
ANALYSIS OF ACTUAL GEOMETRY AND SB LOCA TRANSIENT IN 2-D
GEOliETRIES PROVIDED TEMPERATURE PROFILES IN COLD LEG A D VESSEL
DOWJCOMER
- BENCHIMARKED AGAINST CREARE 1/5 SCALE MODEL DATA WITH GOOD PREDICTION
OF THE I IITED TEMPERATURE DATA
flIXING ANALYSES EMPLOYED IN
BW/OG THERMAL SHOCK ANALYSIS
TEMPERATURE PROFILES IN COLD LEG AND DOWNCOMER
- DETERMINED THAT SIGNIFICANT MIXING OCCURS IN COLD LEG
- QUALIFIED FOR.PTS EVALUATIONS VIA BENCHMARKS WITH
ANALYTICAL SOLUTIONS
- PRESENTLY BEING MODIFIED BY B&WAND INDUSTRY TO OBTAIN MORE
ACCURATE RESULTS
- JET TURBULENT iODEL - TOO CONSERVATIVE; AS SHOWN BY CREARE DATA
AND FLOW-2D AND COMMIX PREDICTIONS
- FLOW 2-D - GOOD AGREEMENT WITH CREARE DATA
- SUPPORTS COMMIX's RESULTS
- COMMIX-1A - PREDICTS RESULTS FOR BWI/ARC TEST DATA WITH GOOD
AGREEMENT
FLOW-2D AND COMMIX-lA PROVIDE REASONABLE TEMPERATURE PROFILES FOR
USE IN PTS ANALYSES
550 - --- - ""-"- 600 SEC.
200
0
0 X
to Turbulent Jet Analysis
7" UNDER NOZZLE FOR CREARE TEST DATA
27.5" UNDER NOZZLE FROM B&W ANALYSIS
.5 I8.8" UNDER NOZZLE FOR CREARE TEST DATA,
34.5" UNDER NOZZLE FROM B&W ANALYSIS
. 4
.3
.2
Distance From Nozzle < Incnes
Figure 4-14. Azimuthal Downcomer Temperature Comparison at 9.2 Inches Below Cold-Leg Centerline for CREARE Test No. 32
155
150
-- FLOW-20
125 I 8.0 10 12 14 16 18 20 22 24 26
Reference distance from not-leg center ine, inches
4-24 Babcock & Wilcox J,-
Figure 4-1. COMmixlA Computational Mesh for TI- Cold Leg and Vessel Downcomer
VENT VRC PUMP DISCHARGE
VENT
VALVES.
90
Figure 4-2. Location of TMI-1 Vessel Welds in.6, Z Coordinates for SBLOCA Transient
- ,,-VENT VALVES IN
PLANE 1=2
4-13
JCOLD LEG PIPNG
0.98
STATION 4 V,D STATION 1 2.8" 3.85" T ,D T
.STATION 5 1.84_ V, T VT 3.78" 0.406* REFERENCE WALL
6.81'
D - DYE INJECTION POINT
DATA
160
150
140
130
120 =%
Distance From Reference Wall (Vertical Leg), in.
4-18
W& NRC
SCREENING CRITERIA
GENERIC APPROACH
- SELECTION OF OVERCOOLING TRANSIENT
- VESSEL THERMAL ANALYSIS
- FRACTURE MECHANICS ANALYSIS
SB LOCA ACS TRANSIENT PRESSURE ANALYSIS HISTORY
OVERCOrCLING REPRESSURI11 ZATION FRACTURE
TRANSIENT RCS FLUID VESSEL MECHANIC VESSEL ANALYSIS PRESS/TEMP MIXING C WALL THERMAL ANALYSIS LIFE HIjS TOR Y ANALYSIS ANALYSI .S (EFPY).
VESSEL IlATERIAL PROPERTIES
BEGAN JULY 1981 - COMPLETED AUGUST 1982
* GENERIC APPROACH
- BAW-1628 (1980) ORIGINAL INDUSTRY T/S WORK
- BAW-1648 (1981)
PROVIDED DATA TO CONSTRUCT REASONABLE AND MECHANISTIC TRANSIENT
BREAK SIZE - LARGE BREAKS RESULT IN COMPLETE DEPRESSURIZING.
SiALL BREAKS RESULT IN HIGHER TEMPERATURES. MlOST PROBABLE
FAILURE IS PORVI HOWEVER, PZR CODE SAFETY VALVE, .023 FT.21 is
IN THE RANGE THAT RESULTS IN COLDEST DOWNCOMER TEMPERATURE 'WITHOUT COMPLETELY DEPRESSURIZING RCS.
BREAK LOCATION - BREAK ON TOP OF PRESSURIZER OR HOT LEGS. BREAK
IN COLD LEG IS LESS SEVERE BECAUSE: 1) HPI FLOWS OUT BREAK,
AND 2) INCREASED VENT VALVE FLOW THROUGH COLD LEG BREAK.
INITIATING EVENT - LOSS OF ALL FW - GIVE MECHANISTIC WAY OF
INITIATING TRANSIENT WHICH LIFT PZR CODE SAFETIES AND INTERRUPT
N/C.
Babcock&
(CONTINUED)
HPI FLOW - MAXIMUM HPI FLOW FOR OCONEE CLASS PLANTS IS 3 HPI PUMPS,
SG HEAT SINK AVAILABILITY - SINCE NATURAL CIRCULATION WAS NOT MAINTAINED, S.G. HEAT SINK IS NOT IMPORTANT,
OP ACTIONJ - TRIP RC PUMPS.
- THROTTLE HPI FLOW TO LIMIT CORE OUTLET SUBCOOLING
TO ~1000F (~-93 N INUTES).
Babcock &Wilco .'~~ti~~Jh.
OVERCOOLI:NG TRANSIENT
- CONSUMERS SENSITIVITY STUDY, DENTON's SHOW CAUSE LETTER
(50.54(8))
- SMUD TRANSIENT (3/20/78); LIGHT BULB INCIDENT
- RESPONSE TO I.E. BULLETIN 79-05C, EFFECTS OF TRIPPING
RC PUMPS ON NON-LOCA TRANSIENT
- SMUD FSAR - MSLB WITH TSV FAILURE IN OPPOSITE LOOP
OPERATING EXPERIENCE - OCONEE DPC SUBMITTAL TO NRC 1/15/82 (DPC-RS-1001)
- TMI-1 GPUN SUBMITTAL TO NRC 6/7/82
THIS INFORMATION PROVIDED GREATER THAN 60 INITIATING EVENTS OR COMB.
OF I.E./MULTI-FAILURES.
TRANSIENT.
- MSLB FOR FPC, AP&L, SMUD.
INITIATING EVENT - ICS SINGLE FAILURE - MECHAiNISTIC AND REALISTIC WAY
OF INITIATING TRANSIENT
Babcock &Wilcox
OVERCOOLING TRANSIENT
(CONTINUED)
AFW SYSTEM - INITIATES, FILLS AT vAX FLOW UNTIL LEVEL SETPOINT IS
REACHED: CONTROLS AT SETPOINT.
- NO THROTTLE OF HPI FOR MSLB TRANSIENT
- THROTTLE HPI AND SHUTOFF FOR TMI-1 TBV TRANSIENT
- 0 THROTTLE OF HPI FOR TBV 0-I TRANSIENT
Babcock Wilcor a Ucetmoott cmt-c.
OPERATING PROCEDURES AND TRAINING.
IN THE PROCESS OF-BEING IMPLEMENTED ON A SCHEDULE
CONSISTENT WITH REQUIREMENTS OF SECY 82-111
* CERTAIN ACTIONS HAVE ALREADY BEEN INCORPORATED
INTO EXISTING PROCEDURES.
* B&W OWNERS GROUP HAS BEEN WORKING CLOSELY WITH THE NRC STAFF IN THE REVIEW AND APPROVAL OF THE
GENERIC GUIDELINES
* GENERIC TRAINING MODULE ON PTS DEVELOPED AND IMPLEMENTED ON UTILITY SPECIFIC SCHEDULES
* NRC PTS AUDIT REVEALED NO SIGNIFICANT PROCEDURAL OR TRAINING DEFICIENCIES
PRESSURIZED THERMAL SHOCK TRAINING OUTLINE
REACTOR VESSEL THERMAL SHOCK DESCRIPTION
FACTORS AFFECTING REACTOR VESSEL THERMAL SHOCK
EFFECTS OF THESE FACTORS
PLANT MODIFICATIONS
COMPLETED/UNDERWAY WHICH
o IMPROVED ICS/NNI POWER SUPPLY RELIABILITY
(ADDRESSES 1978 R-S AND 1980 CR-3 EVENTS)
o UPGRADED Aux. FEEDWATER SYST.EMS
o ANTICIPATORY TRIPs, PORV RELIABILITY IMPROVED
0 IMPROVED SYSTEMS To REDUCE OVERFEED. AND
EXCESS HEAT REMOVAL TRANSIENTS
PHASE SHORT TERM PROGRAM
TAS KA- BEST ESTMIATE DESIGN CURVE.S TASK B REFINEMENT OF EUTRON FLUENCE TASK C - CHARACTERIZATION OF CHEMICAL COMPOSITION TASK D INTEGRATED SURVEI-LLANCE PROGRAM TASK E - COMPLIANCE WITH SECTION V, PARAGRAPH E.,
1OCFR50 APPENDIX G
PHASE II IRRADIATION OF TEST SPECIMENS
PHASE IV -EVALUATION OF IRRADIATED MATERIALS DATA
PHASE FRACTURE MECHANICS ANALYSIS
PHASE VIII - EVALUATION OF ATYPICAL WELD METAL
PHASE IX - DESIGN BASIS FLAW SIZE (ENHANCED ISI)
PHASE X -DOSIMETRY
Pr - I ~ -r~lo
PlI;SrS il-tv PP ASSSIII-IV
DEVELOP IIELD IETAL RESOLVE SitOtT FrA\CTURE TOUGMIESS TERM ISSUE DATA BASE .PHIASES, IX_ A X
RECUCE COurYATIS 'S 4IN DeSIGa CASES
DEVELOP ANALYSIS PROCEDURES
rrE VII
M41 I TOl /EVALUATE DMttSTRATE fttACTOR VESSEL R.Y. A J\ALING INTEGRITY BY ATU1LYSIS
PHASE VI
PAIMEAL ItcsuILi EMPHASIS ON ANALYSIS APPLICATIONS AND PPOCtiES PASED ON
STATE-OF-THE-ART DEVELOPMENTS IH INJDUSTRY-WIDE R&D PcXtPAPS.
BRP O\'ERS GROUP PTS PROGRAM
PER ENCLOSURE A (CHAPTER 9) OF SECY-82-465:
ANALYSIS/ACTIONS REQUIRED WHEN RTNDT SCREENING
CRITERIA ARE EXCEEDED OR WILL EXCEED WITHIN 3 YRS
A. VESSEL MATERIAL PROPERTIES
o REFINEMENT OF CHEMISTRY INFO FOR CRITERIA
MATERIALS
C. FLUX REDUCTION PROGRAMS
ABOVE ARE ADDRESSED BY ONGOING B&W OG INTEGRATED RV
MATERIAL SURVEILLANCE PROGRAM'(IRVMSP)
VESSEL MATERIAL PROPERTIES
MARCH 1976
BAW-1511P, OCTOBER 1980,
OCTOBER 1980, ALSO IN BAW-1485.
PHASE X - DOSIMETRY - METHODS TO FURTHER
REDUCE UNCERTAINTY
o OCONEE-I REPORT
o TMI-1 REPORT
o VESSEL WALL THICKNESS a CLAD THICKNESS: VESSEL INNER RADIUS
o LOCATION & ORIENTATION OF THE ASSUMED
INITIAL CRACK
PROPERTIES, K, E vs. TEMPERATURE
ASSUMED CRACK SHAPE AT INITIATION 8 TIME
OF INITIATION
o TREATMENT OF CLADDING INDUCED STRESSES
o UPPER SHELF TOUGHNESS
RTNDT (AT THE INNER VESSEL RADIUS)
NOTE:
RESULTS OF STAFF REVIEW OF TWO REPORTS SUBMITTED
Peak Reactor Vessel Fluence In Oconee 1
3.0
2.0 RV Surveillance Program
'onversion To Low Leakage Fuel Cycle 0 J
0 4 8 12 16 20 24 28 32
Effective Full Power Years ,(EFPY)
.Lq
' UY U u1KY K)U u U Lm UI Per NRC Procedure
AS OF 12/31/81- GUTHRIE + 2(T
32EFPY GUTHRIE + 2 O
CR-3
OCONEE
.I END
6 WITH ARIS (AUTOMATIC REACTOR INSPECTION SYSTEM), 0-I, O-II, o-III,
AND ANO-I HAVE COMPLETED THEIR 10-YEAR INSPECTION OF THE REACTOR
VESSELS.
* MEETINGS WITH THE NRC STAFF ESTABLISHED THAT THIS PLANNED INSPECTION
OF 0-I WOULD SATISFY THE INTENT OF RG 1.150 BEFORE THE TESTING WAS
BEGUN,
* THE NEAR SURFACE OF ALL BELTLINE WELDS WERE THOROUGHLY INSPECTED.
* WITH ARIS, IT WAS DEMONSTRATED WITH A HIGH DEGREE OF CONFIDENCE THAT
THE BELTLINE REGION OF THE VESSEL IS FREE OF SIGNIFICANT DEFECTS.
* THE INSPECTION RESULTS FOR 0-1, II, AND III ARE:
- NO INDICATION HAD A THRU-WALL DIMENSION GREATER THAN .15 INCHES.
- OF 133 INDICATIONS FOUND IN 0-1 VESSEL, THE VAST MAJORITY (114)
WERE CHARACTERIZED AS LAMINAR INDICATIONS, 3 AS PLANER SUBSURFACE
INDICATIONS AND 16 AS SLAG INCLUSIONS.
- 0-II HAD 4 INDICATIONS.
- 0-III HAD, 1 INDICATION.
- ANO-1 HAD 40 INDICATIONS (PRELIMINARY RESULTS - NOTHING SIGNIFICANT).
* ACCEPTANCE CRITERIA, FROM SECTION XI OF ASME CODE, IS THAT ALL
INDICATIONS BE LESS THAN 1/40 T OR SUPPLEMENTARY ANALYSIS BE PERFORMED
TO DEMONSTRATE THAT SUBSEQUENT OPERATION FOR THAT COMPONENT IS
JUSTIFIED.
TRIP TRANS ENT
PROGRAM TO FOLLOW THIS WORK
o ORNL -SUPPORTING ONGOING EVALUATION OF
0-1 BY ORN
MIXING 1/5 SCALE FACILITY ON B08 PLANTS
PROVIDED INFORMATION FOR CREARE
1/2 SCALE MODEL FACILITY
PER ENCLOSURE A (CHAPTER .9) OF SECY-82-,465:
ANALYSIS/ACTIONS REQUIRED WHEN RTNDT SCREENING
CRITERIA ARE EXCEEDED OR WILL EXCEED WITHIN 3 YRS
A. VESSEL MATERIAL PROPERTIES
o REFINEMENT OF CHEMISTRY INFO FOR CRITERIA
MATERIALS
C. FLUX REDUCTION PROGRAMIS
ABOVE ARE ADDRESSED BY ONGOING B&W OG INTEGRATED RV
MATERIAL SURVEILLANCE PROGRAM.(IRVMSP)
VESSEL MATERIAL PROPERTIES
MARCH 1976
BAW-1511P, OCTOBER 1980.
OCTOBER 1980, ALSO IN BAW-1485,
PHASE X - DOSIMETRY - METHODS TO FURTHER
REDUCE UNCERTAINTY
DETERMINISTIC FRACTURE MECHANICS
o OCONEE-I REPORT
o TMI-1 REPORT
VESSEL INNER RADIUS
PROPERTIES, K, Ed, vs. TEMPERATURE
o ASSUMED CRACK SHAPE AT INITIATION 9 TIME
OF IN ITIATION
o UPPER SHELF TOUGHNESS
RTNDT (AT THE INNER VESSEL RADIUS)
NOTE:
RESULTS OF STAFF REVIEW OF TWO REPORTS SUBMITTED
OTHER PROGRAMS SUPPORTED
BY ONERS GROUP
TRIP TRANSIENT
PROGRAM TO FOLLOW THIS WORK
o'. ORNL - SUPPORTING ONGOING EVALUATION OF
0-I BY ORNL
PROVIDED INFORMATION FOR CREARE
1/2 SCALE MODEL FACLITY
FUTURE ACTIONS SUPPORTING PTS
POSITION ON BW PLANTS
TRANSIENT SELECTION WAS CONSERVATIVE
PREDICTED TO REACH SCREENING CRITERIA
o IMPLEMENT PLANT.MODS AS NEEDED,TO RESPOND TO ANY
NEW ACTUAL PTS EVENTS
0 CONTINUE RV MATERIALS PROGRAM
o EVALUATE FURTHER FLUENCE REDUCTION MEASURES
O CONDUCT 10-YEAR INSERVICE INSPECTIONS OF REACTOR
VESSEL WELDS WITH ENHANCED ISM 1ETHODOLOGY
o CONTINUE TRANSIENT ASSESSMENT PROGRAM
o CONTINUE INVOLVEMENT AND SUPPORT OF INDUSTRY
PROGRAMS
PROGRAM IN PLACE WHICH IS
RESPONSIVE TO THE ITEMS IN
CHAPTER 9 OF STAFF'S REPORT.
o SCREENING CRITERIA IS
PLAN,!TS,
ESSENTIAL TO COMPLETE THE
DEMONSTRATION THAT PTS RISK
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